In a reactor, for example, in a boiling water reactor (BWR), a core performance such as a power distribution and a thermal state of a reactor core are monitored by means of a process control computer included in a reactor power distribution monitor system.
In order to monitor the aforesaid reactor power distribution and thermal state, there is a method of computing a core power distribution with the use of reactor core present data measuring means and a physical model (core three-dimensional nuclear hydrothermal computing code) stored in a process control computer on the basis of the measured reactor core present data and confirming whether a maximum linear heat generation ratio (MLHGR) or a minimum critical power ratio (MCPR) satisfies individual predetermined operation limit value. According to such a method, a reactor operation is carried out.
FIG. 26 and FIG. 27 show a general reactor power distribution monitor system of a boiling water type reactor. In the boiling water type reactor, a reactor pressure vessel 2 is housed in a reactor container 1, and a reactor core 3 is housed in the reactor pressure vessel 2. The reactor core 3 is constructed in a manner that a plurality of fuel assemblies 4 and control rods 5 and the like are mounted. An incore nuclear instrumented fuel assembly 6 is located on a position surrounded by the fuel assemblies 4 of the reactor core 3.
As shown in FIG. 27, a corner gap G formed by four fuel assemblies 4 is provided with an incore nuclear instrumented fuel assembly 6, and a nuclear instrumentation tube 7 is provided with a neutron detector 8 which is dispersively arranged at a plurality of portions in a core axial direction. The neutron detector 8 has a so-called fixed type (stationary or immovable) structure, and in the boiling water reactor, usually, four neutron detectors are dispersively arranged on an effective portion in a fuel axial direction at equal intervals.
Further, the nuclear instrumentation tube 7 is provided with a TIP (Traversing In-Core Probe: movable incore instrumentation) guide tube 9. One movable neutron detector (TIP) 10 is located so as to be movable in an axial direction. As shown in FIG. 26, there is provided a movable type neutron flux measuring system which continuously measures a neutron flux and is movable in an axial direction by means of a retrieval device (selector) 11, a TIP drive unit 12, a TIP drive control device and a TIP neutron flux signal processor 13 or the like. A reference numeral 14 denotes a penetration section, 15 denotes a valve mechanism and 16 denotes a shielding container. These neutron detectors 8 and 10 and their control device such as signal processors 13 and 17 (will be described later) are called as a reactor nuclear instrumentation system 24.
On the other hand, the fixed type (stationary or immovable) neutron detector (LPRM detector) 8 arranged in the reactor core generates an average signal (APRM signal) for each of some divided groups, and then monitors a power level of a power range of the reactor core 3. Further, the fixed type neutron detector 8 constitutes a reactor safety guard system which rapidly makes a scram-operation with respect to a reactor stop system (not shown) such as a control rod drive mechanism in order to prevent a breakdown of a fuel and a reactor when there occurs an abnormal transient phenomenon or accident such that a neutron flux rapidly increases.
By the way, in the fixed type neutron detector 8, a change in sensitivity happens in individual detectors by neutron heat. For this reason, in order to compare and correct the sensitivity of each neutron detector 8 every a predetermined period during operation, the TIP (movable neutron detector) 10 is actuated so as to obtain a continuous power distribution in a core axial direction, and the change in sensitivity of each neutron detector 8 is corrected by a gain adjusting function of a power range detector signal processing unit 17.
A neutron flux signal obtained by the TIP 10 is processed as a neutron flux signal corresponding to a core axial direction position by means of a TIP neutron flux signal processing unit 13 constituting a reactor nuclear instrumentation system 24. Further, in a reactor power distribution computing device 18 (which is usually built in one or plural of process control computers for monitoring an operation of an atomic power generation plant as a program), the neutron flux signal is read as a reference power distribution when computing a three-dimensional hydrothermal force. The reactor power distribution computing device 18 includes a power distribution computing module 19, a power distribution learning module 20 and an input-output unit 21.
Reading a control rod pattern obtained from a present data measuring device 22 which functions as reactor core present data measuring means, a core flow rate, a reactor doom pressure, a reactor heat power obtained from various core present data, and a process data such as a core inlet coolant temperature or the like, these data are processed by means of a present data processing unit 23, and then, are supplied to the reactor power distribution computing unit 18. The present data measuring device 22 is actually composed of a plurality of monitor equipments and is shown as one example of a measuring device for simplification although it is generally named as a device for collecting process data of various operation parameters in the reactor as shown in FIG. 26. Further, the present data processing unit 23 is composed of a process control computer or a part thereof, and a processed core present process data is supplied to the power distribution computing device 18. The power distribution computing module 19 computes a reactor core power distribution according to the three-dimensional nuclear hydrothermal computing code stored in the process control computer, and then, supplies the computed result to the power distribution learning module 20. The power distribution learning module learns on the basis of the reference power distribution, and then, correct the computed result, and thus, accurately computes a reactor power distribution in a power distribution predictive computation after that.
In the conventional incore nuclear instrumented fuel assembly 6, as shown in a perspective view partly in section of FIG. 28, a movable type .gamma.-ray detector 10A may be used in place of the movable neutron detector 10. The movable type .gamma.-ray detector 10A is movable in a core axial direction so as to continuously measure a .gamma.-ray flux in the core axial direction. The .gamma.-ray is generated in proportion to a fission rate in the reactor core 3, and therefore, by measuring a .gamma.-ray flux, it is possible to measure a fission rate in the vicinity of the reactor core.
By using the movable type neutron detector 10 and the movable type .gamma.-ray detector 10A, it is possible to compare and correct a dispersion on detection accuracy in each of the plurality of neutron detectors 8 arranged in the core axial direction and to continuously measure a power distribution in the core axial direction.
As described above, in the conventional reactor nuclear instrumentation system, continuous measurement of the axial direction power distribution of the reactor core 3 depends on the movable type neutron detector 10 and the movable type .gamma.-ray detector 10A which are a movable type measuring device.
Further, there is a conventional reactor nuclear instrumentation system disclosed in Japanese Patent Laid-open Publication No. HEI 6-289182. In the reactor nuclear instrumentation system, a reactor core is provided with an incore nuclear instrumented fuel assembly. The incore nuclear instrumented fuel assembly is constructed in a manner that a fixed type neutron detector assembly and a fixed type gamma thermometer are housed in a nuclear instrumentation tube. The fixed type gamma thermometer is constructed in a manner that many .gamma.-ray heat detectors are dispersively arranged in a core axial direction. These .gamma.-ray heat detectors are arranged at wide intervals in the middle portion of the core axial direction, and are arranged at narrow intervals in an end portion of the core axial direction. The .gamma.-ray heat detector situated on the uppermost end is arranged on a position within 15 cm from the upper end of a fuel effective portion in the core axial direction and measures a .gamma.-ray flux.
In the conventional reactor nuclear instrumentation system, in order to accurately monitor a power distribution in the core axial direction, the movable neutron detector 10 or the movable .gamma.-ray detector 10A is required. For this reason, in the case where only movable neutron detector has been used, there is a problem that it is difficult to monitor a power distribution in the core axial direction with a high accuracy.
In the movable neutron detector 10 or the movable .gamma.-ray detector 10A, at least one neutron detector 10 or .gamma.-ray detector 10A must be vertically moved over a range from an outside of the reactor pressure vessel 2 housing the reactor core 3 to the whole length (core axial length) of the reactor core 3 in the TIP guide tube 9 so as to monitor the power distribution. For this reason, this is a factor of making large a mechanical drive device for moving the neutron detector 10 and the .gamma.-ray detector 10A, and its structure is made complicated, and as a result, there is a problem that a moving operation and maintenance are troublesome. In particular, there are required maintenance and management for mechanical drive devices such as the detector driving device for moving the neutron detector 10 and the .gamma.-ray detector 10A, the retrieval device 11 for selecting the TIP guide tube 9, the valve mechanism 15, the shield container 16 or the like. Further, the movable type detectors 10 and 10A are activated, and for this reason, their maintenance work is a work having the possibility that an worker is exposed.
In view of the above problem, a skilled person is groping a method of monitoring a power distribution in a core axial direction without using a movable measuring device in the reactor nuclear instrumentation system.
The incore nuclear instrumented fuel assembly 6 used in the conventional reactor nuclear instrumentation system is usually provided with four movable neutron detectors 8 and one movable type neutron detector (TIP) 10 or the movable .gamma.-ray detector 10A. Nowadays, a study is made such that a fixed type .gamma.-ray detector in place of the TIP is arranged in the same manner as the fixed type neutron detector 8.
However, in the case where four fixed type .gamma.-ray detectors are arranged in the core axial direction, it is impossible to measure a power on the upper portion and the lower portion of the reactor core 3. Further, in the case of extrapolating a power on the upper portion and the lower portion of the reactor core 3 from four measured data or in the case of interpolating it from four measured data, a behavior in a change of power distribution is different at each portion of the core axial direction. For this reason, a great measurement error is caused, and as a result, an accuracy becomes worse.
Moreover, in the fuel assembly 4 mounted in the reactor core 3 used in a boiling water reactor, in order to keep each interval between fuel rods with a predetermined distance, a plurality of fuel spacers are dispersively located in an axial direction of the fuel assembly 4. In a node where the fuel spacer dispersively exists in the axial direction of the reactor core 3, a neutron flux becomes low due to an elimination effect of a moderator by the fuel spacer, and for this reason, the following matter is anticipated. That is, its power distribution provides a concave power distribution such that a power locally becomes low. However, the three-dimensional nuclear hydrothermal model stored in the conventional process control computer does not deal with the power distribution as described above. For this reason, in the reactor power distribution computing device 18, an error in a power distribution computation in the core axial direction has been corrected by learning a value read by the movable type detector. If the movable detector is replaced with a fixed type detector, an information on correction is not obtained. Thus providing a problem that an error is caused in an evaluation of power on the node where the fuel spacer exists.
Accordingly, in the case where the reactor nuclear instrumentation system is provided with only fixed type measuring device, a measurement error becomes great in a power distribution of the core axial direction. For this reason, there is a need of previously having a freedom of restricting conditions on a reactor operation. As a result, a degree of freedom on a reactor operation is decreased, thus also providing problem of giving an influence to an available factor.
In order to improve an accuracy of measuring a power distribution of the core axial direction, it is considered that many fixed type .gamma.-ray detectors are arranged in the core axial direction. In this case, a detector signal line is increased, and there is a restriction of the number of detector connecting cables which are capable of passing through the nuclear instrumentation tube 9 of the incore nuclear instrumented fuel assembly 6. For this reason, there is a limit to locate many .gamma.-ray detectors.
As disclosed in Japanese Patent Laid-open Publication No. HEI 6-289182, it is considered that the reactor nuclear instrumentation system is provided with many .gamma.-ray heat detectors. However, in the reactor nuclear instrumentation system, there is no knowledge enough to an analysis on a .gamma.-ray heat contributing range and .gamma.-ray heat, and at least one of .gamma.-ray heat detectors located on the upper and lower ends is arranged on a position within a range of 15 cm from the upper and lower ends of a fuel effective portion of the core axial direction. For this reason, it is difficult to accurately detect a .gamma.-ray heat on the upper and lower ends of a fuel effective portion of the core axial direction.